• Complex
  • Title
  • Author
  • Keyword
  • Abstract
  • Scholars
Search
High Impact Results & Cited Count Trend for Year Keyword Cloud and Partner Relationship

Query:

学者姓名:吴宏春

Refining:

Source

Submit Unfold

Co-Author

Submit Unfold

Language

Submit

Clean All

Export Sort by:
Default
  • Default
  • Title
  • Year
  • WOS Cited Count
  • Impact factor
  • Ascending
  • Descending
< Page ,Total 55 >
Method research and engineering application of the B10-abundance correction for PWR EI SCIE
期刊论文 | 2021 , 378 | NUCLEAR ENGINEERING AND DESIGN
Abstract&Keyword Cite

Abstract :

In this paper, the method research and engineering application of the B10-abundance correction for PWR have been implemented. The boric acid is one of the main neutron-absorber materials, within which the B10 is the most significant isotope for absorbing the thermal neutron. Due to the depletion effect and boronation effect, the B10 abundance would variate with the time. These effects would result in the phenomenon that the calculation values of the critical boron concentrations (CBC) provided by nuclear-design code would be smaller than corresponding measurement values. Therefore, for reducing the calculation errors of CBC, the B10-abundance correction method has been proposed based on the Bamboo-C code. Applying the B10-abundance correction method, the engineering validation has been implemented, comparing the calculation values of CBC and B10 abundances with corresponding measurement values. The numerical results indicated that the proposed correction method for B10 abundance can notably improve the calculation accuracy.

Keyword :

B10-abundance correction method Bamboo-C Boronation effect Depletion effect

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Wan, Chenghui , Bai, Jiahe , Liu, Yu et al. Method research and engineering application of the B10-abundance correction for PWR [J]. | NUCLEAR ENGINEERING AND DESIGN , 2021 , 378 .
MLA Wan, Chenghui et al. "Method research and engineering application of the B10-abundance correction for PWR" . | NUCLEAR ENGINEERING AND DESIGN 378 (2021) .
APA Wan, Chenghui , Bai, Jiahe , Liu, Yu , Huang, Xing , Wu, Hongchun . Method research and engineering application of the B10-abundance correction for PWR . | NUCLEAR ENGINEERING AND DESIGN , 2021 , 378 .
Export to NoteExpress RIS BibTex
A new online energy group condensation method for the high-fidelity neutronics code NECP-X EI SCIE
期刊论文 | 2021 , 158 | ANNALS OF NUCLEAR ENERGY
Abstract&Keyword Cite

Abstract :

The high-fidelity whole-core heterogeneous transport calculation attracts a lot of attention because of its high accuracy. However, too many energy-groups for the transport calculation lead to more computational time. This paper presents a new online energy group condensation (OGC) method to accelerate the whole-core heterogeneous transport calculation. Three different models are established for the fuel pin-cell, none-fuel pin-cell and reflector regions, to obtain the spectrum. For the fuel pin-cell regions, a one-dimension (1D) cylinder model is generated during the global-local resonance calculations, and the spectrum is calculated based on the equivalent 1D model and the macro cross-section is condensed based on this spectrum. For the none-fuel pin-cell regions, a super-cell model is established with this none-fuel pin-cell and its surrounding fuel, and the spectrum is calculated with the super-cell model. For the reflector regions, the fuel-lattice-reflector models are generated to calculate spectrum in reflector regions offline, and the spectrum is built-in for condensing the cross-sections for all reflector regions. The condensed energy group structure is studied with the particle swarm optimization (PSO) method to get a superior energy group structure with fewer group numbers but preserving accuracy. A set of benchmark problems are tested and the results show good performance of the new online group condensation method for the transport calculations. (C) 2021 Elsevier Ltd. All rights reserved.

Keyword :

NECP-X Particle swarm optimization The online energy group condensation The whole-core heterogeneous transport calculations

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Zhou, Xinyu , Liu, Zhouyu , Cao, Liangzhi et al. A new online energy group condensation method for the high-fidelity neutronics code NECP-X [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 158 .
MLA Zhou, Xinyu et al. "A new online energy group condensation method for the high-fidelity neutronics code NECP-X" . | ANNALS OF NUCLEAR ENERGY 158 (2021) .
APA Zhou, Xinyu , Liu, Zhouyu , Cao, Liangzhi , Wu, Hongchun , Zhai, Zian . A new online energy group condensation method for the high-fidelity neutronics code NECP-X . | ANNALS OF NUCLEAR ENERGY , 2021 , 158 .
Export to NoteExpress RIS BibTex
Improved block-Jacobi parallel algorithm for the SN nodal method with unstructured mesh EI SCIE
期刊论文 | 2021 , 133 | Progress in Nuclear Energy
WoS CC Cited Count: 1
Abstract&Keyword Cite

Abstract :

The substantial time required for solving the neutron transport equation can be reduced through parallel calculation. The block-Jacobi algorithm is a common method for parallel neutron transport calculation in an unstructured mesh; however, iteration degradation is an inevitable problem that limits the application of this algorithm by reducing the parallel efficiency. In this study, we applied the block-Jacobi algorithm to the SN nodal method with a triangular-z mesh and proposed an improvement to achieve high parallel efficiency. The interface prediction (IP) method was developed to prevent iteration degradation. This method is based on extrapolating the interface information instead of using the information from the preceding iteration at the interfaces in sub-domains. Meanwhile, the prediction was used recursively to accelerate the self-group scattering source iteration in the entire space; this process is referred to as the inner iteration prediction method (IIP). These two methods effectively prevent iteration degradation and reduce time required for self-group scattering source iteration. A more stable and improved parallel performance was thus achieved. © 2021 Elsevier Ltd

Keyword :

Efficiency Forecasting Iterative methods Mesh generation Neutron flux Transport properties

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Qiao, Liang , Zheng, Youqi , Wu, Hongchun et al. Improved block-Jacobi parallel algorithm for the SN nodal method with unstructured mesh [J]. | Progress in Nuclear Energy , 2021 , 133 .
MLA Qiao, Liang et al. "Improved block-Jacobi parallel algorithm for the SN nodal method with unstructured mesh" . | Progress in Nuclear Energy 133 (2021) .
APA Qiao, Liang , Zheng, Youqi , Wu, Hongchun , Wang, Yongping , Du, Xianan . Improved block-Jacobi parallel algorithm for the SN nodal method with unstructured mesh . | Progress in Nuclear Energy , 2021 , 133 .
Export to NoteExpress RIS BibTex
Thermal scattering law data generation for hydrogen bound in zirconium hydride based on the phonon density of states from first-principles calculations EI SCIE
期刊论文 | 2021 , 161 | ANNALS OF NUCLEAR ENERGY
WoS CC Cited Count: 2
Abstract&Keyword Cite

Abstract :

The thermal neutron scattering cross sections of zirconium hydride (ZrHx) is heavily affected by its lattice structure which can be represented by phonon density of states (DOS). In the past decades, many works tried to get parameterized phonon DOS of ZrHx, by fitting certain types of experimental results. In the present work, we adopt the first-principles calculation to obtain the phonon DOS of ZrH(1.5 )in delta phase and ZrH2 in epsilon phase. The theoretical phonon DOS is used to calculate the thermal scattering cross sections which are then used in the neutronics simulations of several TRIGA reactors. The numerical results show that the phonon DOS obtained by first-principles calculations can produce more accurate scattering cross sections, and improve the neutronics results of TRIGA reactors compared with the phonon DOS models applied in ENDF/B and JEFF evaluated nuclear data libraries. (C) 2021 Elsevier Ltd. All rights reserved.

Keyword :

First-principles NECP-Atlas Thermal scattering cross section Thermal scattering law Zirconium hydride

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Zu, Tiejun , Tang, Yongqiang , Wang, Lipeng et al. Thermal scattering law data generation for hydrogen bound in zirconium hydride based on the phonon density of states from first-principles calculations [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 161 .
MLA Zu, Tiejun et al. "Thermal scattering law data generation for hydrogen bound in zirconium hydride based on the phonon density of states from first-principles calculations" . | ANNALS OF NUCLEAR ENERGY 161 (2021) .
APA Zu, Tiejun , Tang, Yongqiang , Wang, Lipeng , Cao, Liangzhi , Wu, Hongchun . Thermal scattering law data generation for hydrogen bound in zirconium hydride based on the phonon density of states from first-principles calculations . | ANNALS OF NUCLEAR ENERGY , 2021 , 161 .
Export to NoteExpress RIS BibTex
Analysis of the two-step source term calculation method with the high-fidelity source term calculation capability EI SCIE
期刊论文 | 2021 , 155 | Annals of Nuclear Energy
WoS CC Cited Count: 1
Abstract&Keyword Cite

Abstract :

The source term calculation capability is implemented in the high-fidelity neutronics code. With this high-fidelity tool, the accuracy of the two-step source term calculation method is analyzed in this work, including decay heat, the activation photon source term, and the neutron source term. The accuracy of NECP-X for the source term calculation is demonstrated through the comparison with SCALE6.1 and STREAM for a set of cases, and the results show good consistency between different codes. The analysis of the two-step source term calculation method is conducted by the control-variable method, and the approximations of generating cross-sections with the 2D lattice calculations and the homogenization process are analyzed for a full-core case. The numerical results show that approximations may result in some difference for the decay heat and the photon source results for the small core, and the difference is investigated with the high-fidelity source term calculation capability. © 2021 Elsevier Ltd

Keyword :

Homogenization method Neutron sources Photons

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Liu, Zhouyu , Wen, Xingjian , Cao, Lu et al. Analysis of the two-step source term calculation method with the high-fidelity source term calculation capability [J]. | Annals of Nuclear Energy , 2021 , 155 .
MLA Liu, Zhouyu et al. "Analysis of the two-step source term calculation method with the high-fidelity source term calculation capability" . | Annals of Nuclear Energy 155 (2021) .
APA Liu, Zhouyu , Wen, Xingjian , Cao, Lu , Cao, Liangzhi , Wu, Hongchun . Analysis of the two-step source term calculation method with the high-fidelity source term calculation capability . | Annals of Nuclear Energy , 2021 , 155 .
Export to NoteExpress RIS BibTex
Research Progresses of Uncertainty Quantification Methods for High Fidelity Numerical Nuclear Reactor EI
期刊论文 | 2021 , 42 (2) , 1-15 | Nuclear Power Engineering
Abstract&Keyword Cite

Abstract :

Numerical reactor based on high fidelity model and method is with the characteristics of high precision and high resolution, but the inherent uncertainty of nuclear data and other parameters will seriously affect the uncertainty of numerical reactor analysis results. Based on the review of the research progresses of numerical reactors and its uncertainty quantification in the world, this paper focuses on the research progresses in NECP Laboratory of Xi'an Jiaotong University in recent years, including the development of one-step high fidelity numerical reactor program NECP-X, the generation of nuclear data covariance database, the uncertainty propagation method based on deterministic method and sampling method, and the uncertainty quantification in transient calculation. An advanced sampling method COST is proposed. Based on the high fidelity numerical reactor program, the uncertainty propagation of covariance of various nuclear parameters in the steady-state and transient analysis of the reactor core is quantified for the first time, which is of great significance for engineering application of numerical reactors. © 2021, Editorial Board of Journal of Nuclear Power Engineering. All right reserved.

Keyword :

Application programs Nuclear reactors Numerical methods Transient analysis Uncertainty analysis

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Cao, Liangzhi , Zou, Xiaoyang , Liu, Zhouyu et al. Research Progresses of Uncertainty Quantification Methods for High Fidelity Numerical Nuclear Reactor [J]. | Nuclear Power Engineering , 2021 , 42 (2) : 1-15 .
MLA Cao, Liangzhi et al. "Research Progresses of Uncertainty Quantification Methods for High Fidelity Numerical Nuclear Reactor" . | Nuclear Power Engineering 42 . 2 (2021) : 1-15 .
APA Cao, Liangzhi , Zou, Xiaoyang , Liu, Zhouyu , Wan, Chenghui , Wu, Hongchun . Research Progresses of Uncertainty Quantification Methods for High Fidelity Numerical Nuclear Reactor . | Nuclear Power Engineering , 2021 , 42 (2) , 1-15 .
Export to NoteExpress RIS BibTex
Multidimensional multiphysics simulations of the supercritical water-cooled fuel rod behaviors based on a new fuel performance code developed on the MOOSE platform EI SCIE
期刊论文 | 2021 , 375 | Nuclear Engineering and Design
WoS CC Cited Count: 1
Abstract&Keyword Cite

Abstract :

Traditional nuclear fuel performance codes usually employ the operator split methods for multiphysics coupling and the quasi-two-dimensional approach (1.5D approach) for geometry modelling. However, the multiphysics behavior of the fuel element is often tightly coupled, and there exist some kinds of non-axisymmetric fuel elements in reactors. In this paper, a new fuel performance code named NECP-CALF has been developed based on the Multiphysics Object-Oriented Simulation Environment (MOOSE). It solves the multiphysics coupled equations using the JFNK method. A new geometrical approach called mixed dimensional approach is implemented in the new code, which allows users to model fuel elements with more flexibility. A UO2-Zr fuel rod is simulated and the results are compared with BISON and CAMPUS to establish the proof-of-concept of this code. Then a small-scale UO2-Zr fuel rod with azimuthally asymmetric cladding temperature is designed and simulated with the mixed dimensional approach to verify the accuracy and demonstrate the performance. Finally, the code is applied to the fuel performance simulation of a supercritical water-cooled fuel rod with flow blockage, and the results show the impact of the flow blockage on the fuel rod performance. © 2021 Elsevier B.V.

Keyword :

Confined flow Fuels Nuclear fuel elements Oxide minerals Uranium dioxide

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Liu, Zhouyu , Xu, Xiaobei , Wu, Hongchun et al. Multidimensional multiphysics simulations of the supercritical water-cooled fuel rod behaviors based on a new fuel performance code developed on the MOOSE platform [J]. | Nuclear Engineering and Design , 2021 , 375 .
MLA Liu, Zhouyu et al. "Multidimensional multiphysics simulations of the supercritical water-cooled fuel rod behaviors based on a new fuel performance code developed on the MOOSE platform" . | Nuclear Engineering and Design 375 (2021) .
APA Liu, Zhouyu , Xu, Xiaobei , Wu, Hongchun , Cao, Liangzhi . Multidimensional multiphysics simulations of the supercritical water-cooled fuel rod behaviors based on a new fuel performance code developed on the MOOSE platform . | Nuclear Engineering and Design , 2021 , 375 .
Export to NoteExpress RIS BibTex
A Source-expansion-based method for transient Pin-Power reconstruction EI SCIE
期刊论文 | 2021 , 164 | ANNALS OF NUCLEAR ENERGY
Abstract&Keyword Cite

Abstract :

In this paper, the transient pin-power reconstruction based on source-expansion method has been proposed. The transient calculation is sufficient to simulate the reactor-physics process in second-time scale, such as the load-dump process and dynamic rod worth measurement. During the transient process, the pin-power distribution is essential for the determination of the power-peak factors, including Fq and FAH. However, with the widely-applied transverse-integrated nodal methods for transient simulation, the detailed pin-averaged results can't be provided directly. To address this problem, the transient pin power reconstruction was proposed. It was derived from the transient fixed-source problem assuming the total-source and the transient fixed-source terms could be respectively expanded by the fourth order Legendre polynomials and the biquadratic Legendre polynomials constructed from the nine-node problem. Moreover, to improve the accuracy of pin-averaged results, the corner-point condition and corner-discontinuity factors were employed in the reconstruction process. This proposed transient pin power reconstruction method has been implemented in our self-developed code named SPARK and verified with several benchmarks. Results of the verification indicated that this proposed method for transient pin-power reconstruction can provide a satisfactory accuracy in transient process. (C) 2021 Elsevier Ltd. All rights reserved.

Keyword :

Source-expansion method The SPARK code Transient pin-power reconstruction Transient simulation

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Bai, Jiahe , Wan, Chenghui , Li, Yunzhao et al. A Source-expansion-based method for transient Pin-Power reconstruction [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 164 .
MLA Bai, Jiahe et al. "A Source-expansion-based method for transient Pin-Power reconstruction" . | ANNALS OF NUCLEAR ENERGY 164 (2021) .
APA Bai, Jiahe , Wan, Chenghui , Li, Yunzhao , Wu, Hongchun , Li, Fan . A Source-expansion-based method for transient Pin-Power reconstruction . | ANNALS OF NUCLEAR ENERGY , 2021 , 164 .
Export to NoteExpress RIS BibTex
On the equivalence of reaction rate in energy collapsing of fast reactor code SARAX SCIE
期刊论文 | 2021 , 53 (3) , 732-740 | NUCLEAR ENGINEERING AND TECHNOLOGY
WoS CC Cited Count: 1
Abstract&Keyword Cite

Abstract :

Scattering resonance of medium mass nuclides leads complex spectrum in the fast reactor, which requires thousands of energy groups in the spectrum calculation. When the broad-group cross sections are collapsed, reaction rate cannot be completely conserved. To eliminate the error from energy collapsing, the Super-homogenization method in energy collapsing (ESPH) was employed in the fast reactor code SARAX. An ESPH factor was derived based on the ESPH-corrected SN transport equation. By applying the factor in problems with reflective boundary condition, both the effective multiplication factor and reaction rate were conserved. The fixed-source iteration was used to ensure the stability of ESPH iteration. However, in the energy collapsing process of SARAX, the vacuum boundary condition was adopted, which was necessary for fast reactors with strong heterogeneity. To further reduce the error caused by leakage, an additional conservation factor was proposed to correct the neutron current in energy collapsing. To evaluate the performance of ESPH with conservation factor, numerical benchmarks of fast reactors were calculated. The results of broad-group calculation agreed well with the direct full-core Monte-Carlo calculation, including the effective multiplication factor, radial power distribution, total control rod worth and sodium void worth. ? 2020 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).

Keyword :

Conservation factor Energy collapsing Fast reactor Super-homogenization

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Xiao, Bowen , Wei, Linfang , Zheng, Youqi et al. On the equivalence of reaction rate in energy collapsing of fast reactor code SARAX [J]. | NUCLEAR ENGINEERING AND TECHNOLOGY , 2021 , 53 (3) : 732-740 .
MLA Xiao, Bowen et al. "On the equivalence of reaction rate in energy collapsing of fast reactor code SARAX" . | NUCLEAR ENGINEERING AND TECHNOLOGY 53 . 3 (2021) : 732-740 .
APA Xiao, Bowen , Wei, Linfang , Zheng, Youqi , Zhang, Bin , Wu, Hongchun . On the equivalence of reaction rate in energy collapsing of fast reactor code SARAX . | NUCLEAR ENGINEERING AND TECHNOLOGY , 2021 , 53 (3) , 732-740 .
Export to NoteExpress RIS BibTex
Application of CENDL-3.2 and ENDF/B-VIII.0 on the reactor physics simulation of PWR EI SCIE
期刊论文 | 2021 , 158 | ANNALS OF NUCLEAR ENERGY
WoS CC Cited Count: 2
Abstract&Keyword Cite

Abstract :

The latest CENDL and ENDF/B evaluated nuclear data libraries was released in 2020 and 2018, respectively. To apply CENDL-3.2 and ENDF/B-VIII.0 in the reactor physic simulations of pressurized water reactor (PWR), the CNP-1000 PWR, which is an improved GEN-II PWR and operated in Fuqing Nuclear Power Plant in China, is simulated using these two libraries. The key parameters during the startup physics tests and power operation in the first three fuel cycles of the CNP-1000 reactor have been simulated and compared with corresponding measurement values. The numerical results show that ENDF/B-VIII.0 performs better in several parameters of the startup physics tests than ENDF/B-VII.0; the cross-section data in CENDL-3.2 is competent in the engineering application of PWR. (C) 2021 Elsevier Ltd. All rights reserved.

Keyword :

CENDL-3.2 CNP-1000 ENDF/B-VIII.0 LOCUST/SPARK NECP-Atlas

Cite:

Copy from the list or Export to your reference management。

GB/T 7714 Zu, Tiejun , Huang, Yihan , Teng, Qichen et al. Application of CENDL-3.2 and ENDF/B-VIII.0 on the reactor physics simulation of PWR [J]. | ANNALS OF NUCLEAR ENERGY , 2021 , 158 .
MLA Zu, Tiejun et al. "Application of CENDL-3.2 and ENDF/B-VIII.0 on the reactor physics simulation of PWR" . | ANNALS OF NUCLEAR ENERGY 158 (2021) .
APA Zu, Tiejun , Huang, Yihan , Teng, Qichen , Han, Fenglin , Huang, Xing , Wan, Chenghui et al. Application of CENDL-3.2 and ENDF/B-VIII.0 on the reactor physics simulation of PWR . | ANNALS OF NUCLEAR ENERGY , 2021 , 158 .
Export to NoteExpress RIS BibTex
10| 20| 50 per page
< Page ,Total 55 >

Export

Results:

Selected

to

Format:
FAQ| About| Online/Total:692/171459737
Address:XI'AN JIAOTONG UNIVERSITY LIBRARY(No.28, Xianning West Road, Xi'an, Shaanxi Post Code:710049) Contact Us:029-82667865
Copyright:XI'AN JIAOTONG UNIVERSITY LIBRARY Technical Support:Beijing Aegean Software Co., Ltd.